Appendix B. Aging Management Programs and Activities

Appendix B Aging Management Programs and Activities

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TABLE OF CONTENTS Table of Contents...i Appendix B: Aging Management Programs and Activities... B.1-1 B.1 INTRODUCTION... B.1-1 B.2 PROGRAM AND ACTIVITY ATTRIBUTES... B.2-1 B.2.1 TYPES OF PROGRAMS AND ACTIVITIES... B.2-1 B.2.2 ATTRIBUTE DEFINITIONS... B.2-2 B.3 AGING MANAGEMENT PROGRAMS AND ACTIVITIES... B.2.2-1 B.3.1 ALLOY 600 AGING MANAGEMENT REVIEW... B.3.1-1 B.3.2 BATTERY RACK INSPECTIONS... B.3.2-1 B.3.3 BORAFLEX MONITORING PROGRAM... B.3.3-1 B.3.4 BORATED WATER SYSTEMS STAINLESS STEEL INSPECTION... B.3.4-1 B.3.5 BOTTOM-MOUNTED INSTRUMENTATION THIMBLE TUBE INSPECTION PROGRAM... B.3.5-1 B.3.6 CHEMISTRY CONTROL PROGRAM... B.3.6-1 B.3.7 CONTAINMENT INSERVICE INSPECTION PLAN IWE... B.3.7-1 B.3.8 CONTAINMENT LEAK RATE TESTING PROGRAM... B.3.8-1 B.3.9 CONTROL ROD DRIVE MECHANISM NOZZLE AND OTHER VESSEL CLOSURE PENETRATIONS INSPECTION PROGRAM... B.3.9-1 B.3.10 CRANE INSPECTION PROGRAM... B.3.10-1 B.3.11 DIVIDER BARRIER SEAL INSPECTION AND TESTING PROGRAM... B.3.11-1 B.3.12 FIRE PROTECTION PROGRAM... B.3.12-1 B.3.12.1 Fire Barrier Inspections... B.3.12-1 B.3.12.2 Mechanical Fire Protection Component Tests and Inspections... B.3.12-3 B.3.13 FLOOD BARRIER INSPECTION... B.3.13-1 B.3.14 FLOW ACCELERATED CORROSION PROGRAM... B.3.14-1 B.3.15 FLUID LEAK MANAGEMENT PROGRAM... B.3.15-1 B.3.16 GALVANIC SUSCEPTIBILITY INSPECTION... B.3.16-1 B.3.17 HEAT EXCHANGER ACTIVITIES... B.3.17-1 B.3.17.1 Component Cooling Heat Exchangers... B.3.17-1 B.3.17.2 Containment Spray Heat Exchangers... B.3.17-6 B.3.17.3 Diesel Generator Engine Cooling Water Heat Exchangers... B.3.17-9 B.3.17.4 Heat Exchanger Preventive Maintenance Activities Control Area Chilled Water B.3.17-14 B i

B.3.17.5 Heat Exchanger Preventive Maintenance Activities Diesel Generator Engine Starting Air..B.3.17-16 B.3.17.6 Heat Exchanger Preventive Maintenance Activities- Pump Motor Air Handling Units...B.3.17-18 B.3.17.7 Heat Exchanger Preventive Maintenance Activities- Pump Oil Coolers...B.3.17-20 B.3.18 ICE CONDENSER INSPECTIONS...B.3.18-1 B.3.18.1 Ice Basket Inspection...B.3.18-1 B.3.18.2 Ice Condenser Engineering Inspection...B.3.18-3 B.3.19 INACCESSIBLE NON-EQ MEDIUM-VOLTAGE CABLES AGING MANAGEMENT PROGRAM...B.3.19-1 B.3.20 INSERVICE INSPECTION PLAN...B.3.20-1 B.3.20.1 ASME Section XI, Subsections IWB and IWC Inspections...B.3.20-2 B.3.20.2 ASME Section XI, Subsection IWF Inspections...B.3.20-7 B.3.21 INSPECTION PROGRAM FOR CIVIL ENGINEERING STRUCTURES AND COMPONENTS...B.3.21-1 B.3.22 LIQUID WASTE SYSTEM INSPECTION...B.3.22-1 B.3.23 NON-EQ INSULATED CABLES AND CONNECTIONS AGING MANAGEMENT PROGRAM...B.3.23-1 B.3.24 PREVENTIVE MAINTENANCE ACTIVITIES...B.3.24-1 B.3.24.1 Condenser Circulating Water System Internal Coating Inspection...B.3.24-1 B.3.24.2 Refueling Water Storage Tank Internal Coating Inspection...B.3.24-5 B.3.25 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE MONITORING PROGRAM...B.3.25-1 B.3.26 REACTOR VESSEL INTEGRITY PROGRAM...B.3.26-1 B.3.27 REACTOR VESSEL INTERNALS INSPECTION...B.3.27-1 B.3.28 SELECTIVE LEACHING INSPECTION...B.3.28-1 B.3.29 SERVICE WATER PIPING CORROSION PROGRAM...B.3.29-1 B.3.30 STANDBY NUCLEAR SERVICE WATER POND DAM INSPECTION...B.3.30-1 B.3.31 STEAM GENERATOR SURVEILLANCE PROGRAM...B.3.31-1 B.3.32 SUMP PUMP SYSTEMS INSPECTION...B.3.32-1 B.3.33 TECHNICAL SPECIFICATION SR 3.6.16.3 VISUAL INSPECTION...B.3.33-1 B.3.34 TREATED WATER SYSTEMS STAINLESS STEEL INSPECTION...B.3.34-1 B.3.35 UNDERWATER INSPECTION OF NUCLEAR SERVICE WATER STRUCTURES...B.3.35-1 B.3.36 WASTE GAS SYSTEM INSPECTION...B.3.36-1 B.4 REFERENCES FOR APPENDIX B...B.4-1 B - ii

APPENDIX B: AGING MANAGEMENT PROGRAMS AND ACTIVITIES B.1 INTRODUCTION Aging management programs and activities that are credited during the aging management review are described in the remaining sections of Appendix B. The demonstrations, along with the program and activity descriptions, meet the requirement specified in 54.21(a)(3). Along with the technical information contained in Chapters 2, 3, and 4, Appendix B is designed to allow the NRC to make the finding contained in 54.29(a)(1). 54.29 Standards for issuance of a renewed license A renewed license may be issued by the Commission up to the full term authorized by 54.31 if the Commission finds that: (a) Actions have been identified and have been or will be taken with respect to the matters identified in Paragraphs (a)(1) and (a)(2) of this section, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the CLB, and that any changes made to the plant s CLB in order to comply with this paragraph are in accord with the Act and the Commission s regulations. These matters are: (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 54.21(a)(1); and (2) time-limited aging analyses that have been identified to require review under 54.21(c). (b) Any applicable requirements of Subpart A of 10 CFR Part 51 have been satisfied. (c) Any matters raised under 2.758 have been addressed. The aging management programs described in Appendix B include existing, ongoing programs as well as new programs that are currently not implemented. These descriptions of programs and activities are intended to provide an overview of the range of actions required to manage aging. Some of the descriptions have used a series of specific attributes to facilitate the description of the actions. These attributes are defined in Section B.2, of Appendix B. B.1-1

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B.2 PROGRAM AND ACTIVITY ATTRIBUTES Attributes that are utilized in most of the program and activity descriptions for license renewal, with a few exceptions, are described in Appendix B.2. The following information sources served as primary inputs to the attribute definitions used in Appendix B: 1. Application for Renewed Operating Licenses, Oconee Nuclear Station [Reference B - 1] 2. NEI 95-10, Revision 2, Sections 4.2 and 4.3 [Reference B - 2] 3. Draft Standard Review Plan for License Renewal, Appendix A.1 [Reference B - 3] 4. NEI letter dated October 13, 2000 to U. S. Nuclear Regulatory Commission [Reference B - 4] B.2.1 TYPES OF PROGRAMS AND ACTIVITIES The attributes described in the Section B.2.2 are applicable to the following types of programs and activities: One-time Inspections A one-time inspection is performed for components when the presence of aging effects requiring management are not indicated but cannot be ruled out. The inspection is performed one time only and inspects components for specific indications that are linked to degradation caused by specific aging effects. No actions are taken as part of a one-time inspection to prevent the aging effect or trend inspection results. For McGuire, these new inspections will be completed following issuance of renewed operating licenses for McGuire Nuclear Station and by June 12, 2021 (the end of the initial license of McGuire Unit 1). For Catawba, these new inspections will be completed following issuance of the renewed operating licenses for Catawba Nuclear Station and by December 6, 2024 (the end of the initial license of Catawba Unit 1). Completion of these new inspections prior to the end of the initial licenses is consistent with conclusions previously made by the Nuclear Regulatory Commission in NUREG-1723. Prevention Programs The actions of a prevention program preclude specific aging effects from occurring. For example, a coating precludes corrosion of the base metal from occurring. Mitigation Programs A mitigation program attempts to slow the effects of aging. For example, water chemistry mitigates internal corrosion of piping. B.2-1

Condition Monitoring Programs A condition monitoring program inspects or examines the presence or extent of aging effects. For example, the ASME Section XI Inservice Inspection Program which requires visual, surface and volumetric examinations is a condition monitoring program. Performance Monitoring Programs Performance monitoring programs test the ability of a structure or component to perform its intended function. For example, heat balances test the heat transfer function of heat exchanger tubes. B.2.2 ATTRIBUTE DEFINITIONS The attribute definitions used to describe new and existing programs and activities are provided below. Scope This program attribute identifies the specific structures or components managed by the program or activity. Preventive Actions This program attribute describes the actions taken in the period of extended operation to either prevent aging effects from occurring or mitigate (i.e., lessen or slow down) aging degradation for prevention and mitigation programs. This attribute is not applicable for one-time inspections, condition monitoring and performance monitoring programs. Parameters Monitored or Inspected This program attribute describes what is being monitored or inspected for all inspections and programs. These descriptions include the observable parameters or indicators to be monitored or inspected for each aging effect managed. The observable parameters should be linked to the degradation of the structure or component intended functions in the period of extended operation. Detection of Aging Effects The detection of aging effects should occur before there is a loss of structure and component intended function(s). Monitoring & Trending This program attribute describes when, where and how program data is collected; i.e., all aspects of activities to collect data as part of the program. This description includes aspects such as method or technique (e.g., visual, volumetric, surface inspection), frequency, sample size, and timing of new/one-time inspections. This attribute also provides information that links the parameters to be monitored or inspected to the aging effects being managed. Trending is a comparison of the current monitoring results with previous monitoring results in order to make predictions for the future and to initiate actions as necessary. B.2-2

Acceptance Criteria This program attribute describes the acceptance criteria for ensuring the structure or component intended function is maintained during the period of extended operation. The acceptance criteria may be based on design or current licensing basis information as well as established industry codes or standards. Corrective Action & Confirmation Process This program attribute describes the actions to be taken in the extended period of operation when the acceptance criteria or standard is not met. The corrective action and confirmation process that is described for each aging management program or activity applies to all structures and components within the scope of the program or activity. In some cases the program itself includes its own corrective action and confirmation process. In other cases, the corrective action process is credited for corrective action and confirmation process. The corrective action process is a formal corrective action program which facilitates the correction of conditions adverse to quality. Corrective actions are documented. Data are periodically reviewed to identify positive or negative changes and to initiate additional actions, as necessary. The corrective action process is implemented by Nuclear System Directives NSD 208, Problem Investigation Process and NSD 223, Trending of PIP Data. Administrative Controls This program attribute describes the administrative structure under which the programs and activities are executed. Examples of various administrative structures include program manuals, nuclear station directives, engineering support documents, plant procedures, and work orders. The administrative controls provide for a review and approval process. Operating Experience This program attribute provides the objective evidence that supports the determination that the program or activity provides reasonable assurance that the effects of aging will be adequately managed such that the structure or component intended function(s) will be maintained consistent with the current licensing basis during the period of extended operation (i.e., 20-years from the end of the initial operating license). Plant specific operating experience includes licensee event reports, reports documenting the results of the credited program or activity, as well as plant maintenance and operating records. Several programs and activities contained in Appendix B are equivalent to the corresponding programs and activities that were described in the Application for Renewed Operating Licenses of Oconee Nuclear Station, Units 1, 2, and 3. In these instances, the pertinent section of NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, is referenced [Reference B - 5]. B.2-3

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B.3 AGING MANAGEMENT PROGRAMS AND ACTIVITIES B.3.1 ALLOY 600 AGING MANAGEMENT REVIEW Note: The ALLOY 600 AGING MANAGEMENT REVIEW is generically applicable to both McGuire Nuclear Station and Catawba Nuclear Station, except as otherwise noted. The purpose of the Alloy 600 Aging Management Review is to ensure that nickel-based alloy locations are adequately inspected by the Inservice Inspection Plan (Appendix B.3.20) or other existing programs such as the Control Rod Drive Mechanism and Other Vessel Head Penetration Program (Appendix B.3.9), the Reactor Vessel Internals Inspection (Appendix B.3.27), and the Steam Generator Integrity Program (Appendix B.3.31). The review will demonstrate the general oversight and management of cracking due to primary water stress corrosion cracking (PWSCC). The Alloy 600 Aging Management Review will identify Alloy 600/690, 82/182 and 52/152 locations. A ranking of susceptibility to PWSCC will be performed for the nickel-based alloy locations. A review will be performed to ensure that nickel-based alloy locations are adequately inspected by the Inservice Inspection Plan (Appendix B.3.20) or other existing programs such as the Control Rod Drive Mechanism and Other Vessel Head Penetration Program (Appendix B.3.9), the Reactor Vessel Internals Inspection (Appendix B.3.27), and the Steam Generator Integrity Program (Appendix B.3.31). This review will utilize industry and Duke specific operating experience. Inspection method and frequency of inspection for the Alloy 600/690, 82/182, and 52/152 locations for the period of extended operation will be adjusted as needed based on the results of this review. In addition, supplemental inspections for the period of extended operation will be developed as needed. For McGuire, this review will be completed following issuance of renewed operating licenses for McGuire Nuclear Station and by June 12, 2021 (the end of the initial license of McGuire Unit 1). For Catawba, this review will be completed following issuance of renewed operating licenses for Catawba Nuclear Station and by December 6, 2024 (the end of the initial license of Catawba Unit 1). The results of this review will be incorporated into the unit specific inservice inspection (ISI) plans for the ISI intervals during the period of extended operation. B.3.1-1

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B.3.2 BATTERY RACK INSPECTIONS Note: The BATTERY RACK INSPECTIONS are generically applicable to both McGuire Nuclear Station and Catawba Nuclear Station, except as otherwise noted. Loss of material due to corrosion is an aging effect requiring programmatic management for steel battery racks for the extended period of operation. The Battery Rack Inspections are credited with managing loss of material that could impact the intended function of structural support. The Battery Rack Inspections are condition monitoring programs. The regulatory basis for inspecting battery racks is found in the McGuire and Catawba Technical Specifications and Selected Licensee Commitments as identified: McGuire EPL System - Technical Specification (SR) 3.8.4.3 EPQ System - Selected Licensee Commitment 16.8.3.3 EQD System - Selected Licensee Commitment 16.9.7.12 ETM System - Selected Licensee Commitment 16.9.7.17 Catawba EPL System - Technical Specification (SR) 3.8.4.4 EPQ System - Technical Specification (SR) 3.8.4.4 EQD System - Selected Licensee Commitment 16.7-9.2 ETM System - Selected Licensee Commitment 16.7-9.4 Scope The scope of the Battery Rack Inspections include the battery racks for the following systems: EPL System (Vital Batteries) EPQ System (Diesel Generator Batteries) ETM System (Standby Shutdown Facility Batteries) EQD System (Standby Shutdown Facility Diesel Batteries) Preventive Actions No actions are taken as part of this program to prevent aging effects or mitigate aging degradation. Parameters Monitored or Inspected The Battery Rack Inspections provide visual examination of the battery racks for physical damage or abnormal deterioration, including loss of material. B.3.2-1

Detection of Aging Effects In accordance with information provided in Monitoring & Trending, the Battery Rack Inspections will detect loss of material prior to loss of battery rack intended function. Monitoring & Trending The Battery Rack Inspections perform visual inspections to detect loss of material in accordance with McGuire and Catawba Technical Specifications and Selected Licensee Commitments. The inspections are based on guidance provided in IEEE 450-1980 [Reference B - 6]. No actions are taken as part of this program to trend inspection results. EPL System battery racks are inspected every 18 months in accordance with Technical Specifications. EPQ System battery racks are inspected every 18 months in accordance with McGuire Selected Licensee Commitment and Catawba Technical Specification. EQD System battery racks are inspected every 18 months in accordance with McGuire and Catawba Selected Licensee Commitments. ETM System battery racks are inspected every 18 months in accordance with McGuire and Catawba Selected Licensee Commitments. Acceptance Criteria The acceptance criterion is no visual indication of loss of material. Corrective Action & Confirmation Process Areas which do not meet the acceptance criteria are evaluated by the accountable engineer for continued service and repaired as required. Structures and components that are deemed unacceptable are documented under the corrective action program. Specific corrective actions and confirmatory actions, as needed, are implemented in accordance with the corrective action program. Administrative Controls The Battery Rack Inspections are governed by McGuire and Catawba Technical Specifications and Selected Licensee Commitments. The Battery Rack Inspections are implemented by controlled plant procedures, as required by Technical Specification 5.4, and work management system using model work orders. Operating Experience A review of McGuire and Catawba-specific surveillance records did not identify any instances where abnormal deterioration, which would include loss of material, of the battery racks had occurred. B.3.2-2

Conclusion The Battery Rack Inspections have been demonstrated to be capable of detecting and managing loss of material. The Battery Rack Inspections described above are equivalent to the Battery Rack Inspections described and evaluated in NUREG-1723, Section 3.8.3.2.1 [Reference B - 5]. Based on the above review, the continued implementation of the Battery Rack Inspections provides reasonable assurance that loss of material will be managed such that the intended functions of the battery racks will continue to be maintained consistent with the current licensing basis for the period of extended operation (i.e., 20-years from the end of the initial operating license). B.3.2-3

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B.3.3 BORAFLEX MONITORING PROGRAM Note: The BORAFLEX MONITORING PROGRAM is applicable only to McGuire Nuclear Station. Degradation due to gamma irradiation has been identified as an aging effect requiring programmatic management for the Boraflex neutron-absorbing panels in spent fuel storage racks for the extended period of operation. The function of the Boraflex panels is to ensure that reactivity of the storage fuel assemblies is maintained within required limits. Boraflex has been shown to degrade as a result of gamma irradiation and exposure to the spent fuel pool environment. The B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks. The Boraflex Monitoring Program is credited with managing aging of Boraflex panels for the period of extended operation. The Boraflex Monitoring Program is a performance monitoring program. Scope The scope of the Boraflex Monitoring Program includes all Boraflex neutronabsorbing panels in the McGuire Units 1 and 2 spent fuel storage racks. Catawba Nuclear Station does not use Boraflex. Preventive Actions No actions are taken as part of this program to prevent aging effects or mitigate aging degradation. Parameters Monitored or Inspected The Boraflex Monitoring Program monitors the Boraflex panel average storage rack poison material Boron 10 areal density. The panel average Boron 10 areal density is used as an input to the spent fuel pool storage rack criticality calculations. In addition, the silica levels are monitored in the spent fuel pool. The silica levels provide an indication of the depletion of boron carbide from Boraflex. Detection of Aging Effects In accordance with information provided in Monitoring & Trending, the Boraflex Monitoring Program will monitor Boraflex panel areal density prior to loss of intended function. Monitoring & Trending The Boraflex Monitoring Program includes in-situ testing of the Boron 10 areal density. The frequency of testing is every three years. Testing may be performed more frequently based on engineering judgment, spent fuel pool water chemistry, and modeling projections of Boraflex degradation. Selection of Boraflex panels for in-situ testing is based upon predicted Boron 10 areal density loss. Acceptance Criteria The acceptance criteria are based on maintaining the minimum areal density of B4C assumed in the criticality calculations. The requirements are listed in B.3.3-1

McGuire Selected Licensee Commitment (SLC) 16.9.24, Spent Fuel Pool Storage Rack Poison Material. Corrective Action & Confirmation Process The specified corrective actions are identified in McGuire Selected Licensee Commitments 16.9.24, Spent Fuel Pool Storage Rack Poison Material. Structures and components that do not meet the acceptance criteria are evaluated by engineering for continued service and repaired as required. Structures and components which are deemed unacceptable are documented under the corrective action program. Specific corrective actions and confirmatory actions are implemented in accordance with the corrective action program Administrative Controls The Boraflex Monitoring Program is governed by McGuire Selected Licensee Commitment (SLC) 16.9.24. Operating Experience Blackness testing was performed at McGuire on the spent fuel storage racks in 1991. The testing was performed to measure the amount of pullback at the ends of the Boraflex panels caused by shrinkage, as well as the size and frequency of gap formation. Shrinkage and gap formation were observed in both pools at McGuire. The data was incorporated into revised criticality analyses for the storage racks and k eff was verified less than or equal to 0.95 [Reference B - 7]. In 1996 as a result of industry-wide experience with degradation of Boraflex, the NRC issued Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks [Reference B - 8]. Generic Letter 96-04 provided descriptions of several industry experiences and a discussion of relevant experimental data from test programs. The staff stated that on the basis of test and surveillance information from plants that had detected areas of Boraflex degradation, no safety concern existed that warranted immediate action. In issuing Generic Letter 96-04, the staff requested that all licensees with installed spent fuel pool storage racks containing the neutron absorber Boraflex provide an assessment of the physical condition of the Boraflex. The Duke response to this request was provided in a letter to the NRC dated October 22, 1996 [Reference B - 7] and supplemented in December 22, 1997 [Reference B - 9]. The response indicated, in part, that Duke had acquired the RACKLIFE computer code which had been developed by the Electric Power Research Institute for the purpose of assessing overall Boraflex thinning based upon cumulative gamma exposure, storage rack design parameters, and dissolved silica concentration in the spent fuel pool. In-situ measurements were performed that verified that the Boraflex Monitoring Program accurately predicts the Boron 10 areal density. B.3.3-2

Conclusion The Boraflex Monitoring Program has been demonstrated to be capable of detecting and managing degradation of the Boraflex panels. The Boraflex Monitoring Program described above is equivalent to the Boraflex Monitoring Program described and evaluated in NUREG-1723, Section 4.2.10 [Reference B - 5]. Based on the above review, the continued implementation of the Boraflex Monitoring Program provides reasonable assurance that degradation of the Boraflex panels will be managed such that the intended function of the Boraflex panels will continue to be maintained consistent with the current licensing basis for the period of extended operation (i.e., 20-years from the end of the initial operating license). B.3.3-3

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B.3.4 BORATED WATER SYSTEMS STAINLESS STEEL INSPECTION Note: The BORATED WATER SYSTEMS STAINLESS STEEL INSPECTION is generically applicable to both McGuire Nuclear Station and Catawba Nuclear Station, except as otherwise noted. The purpose of the Borated Water Systems Stainless Steel Inspection is to characterize any loss of material or cracking of stainless steel components exposed to alternate wetting and drying in borated water environments. Uncertainty exists as to whether alternate wetting and drying of the borated water could cause aging in stainless steel components such that they may lose their pressure boundary function in the period of extended operation. This activity will inspect stainless steel components exposed to an alternate wetting and drying borated water environment to detect the presence and extent of any loss of material or cracking. The Borated Water Systems Stainless Steel Inspection is a one-time inspection. Scope The scope of the Borated Water Systems Stainless Steel Inspection is stainless steel components exposed to an alternate wetting and drying borated water environment in the following McGuire and Catawba systems: Containment Spray Refueling Water Preventive Actions No actions are taken as part of this program to prevent aging effects or to mitigate aging degradation. Parameters Monitored or Inspected The parameters inspected by the Borated Water Systems Stainless Steel Inspection are pipe wall thickness, as a measure of loss of material, and evidence of cracking. Detection of Aging Effects The Borated Water Systems Stainless Steel Inspection is a one-time inspection that will detect the presence and extent of loss of material or cracking of stainless steel components. Monitoring & Trending The Borated Water Systems Stainless Steel Inspection will inspect stainless steel components, welds, and heat affected zones, as applicable, in the Containment Spray System in the area of the internal air/water interface. The borated water environment found downstream of valves NS-12, 15, 29, 32, 38, and 43 in the Containment Spray System at McGuire and Catawba is stagnant and isolated from the remainder of the system, and therefore, not controlled by the Chemistry Control Program. Water from the refueling water storage tank is introduced during valve testing with level in the piping reaching the same elevation as the tank. Since the pipe is open to containment, evaporation occurs and concentration of contaminants could occur at the air/water interface. This concentration of B.3.4-1

contaminants could lead to loss of material or cracking. Therefore, a one-time inspection around this water line is warranted. One of twelve possible locations at each site will be inspected using volumetric technique. If no parameters are known that would distinguish the susceptible locations at each site, one of the twelve available at each site will be examined based on accessibility and radiological concerns. The results of this inspection will be applied to the specific stainless steel components exposed to an alternate wetting and drying borated water environment in the Refueling Water System. For McGuire, this new inspection will be completed following issuance of renewed operating licenses for McGuire Nuclear Station and by June 12, 2021 (the end of the initial license of McGuire Unit 1). For Catawba, this new inspection will be completed following issuance of renewed operating licenses for Catawba Nuclear Station and by December 6, 2024 (the end of the initial license of Catawba Unit 1). No actions are taken as part of this activity to trend inspection results. Should industry data or other evaluations indicate that the above inspections can be modified or eliminated, Duke will provide plant-specific justification to demonstrate the basis for the modification or elimination. Acceptance Criteria The acceptance criteria for the Borated Water Systems Stainless Steel Inspection is no unacceptable loss of material or cracking that could result in a loss of the component intended function(s) as determined by engineering evaluation. Corrective Action & Confirmation Process If engineering evaluation determines that continuation of the aging effects will not cause a loss of component intended function(s) under any current licensing basis design conditions for the period of extended operation, then the aging management review is complete and no further action is required. If engineering evaluation determines that additional information is required to more fully characterize any or all of the aging effects, then additional inspections will be completed or other actions taken in order to obtain the additional information. If further engineering evaluation determines that continuation of the aging effects could cause a loss of component intended function(s) under current licensing basis design conditions for the period of extended operation, then programmatic oversight will be defined. Specific corrective actions will be implemented in accordance with the corrective action program. Administrative Controls The Borated Water Systems Stainless Steel Inspection will be implemented in accordance with controlled plant procedures. B.3.4-2

Operating Experience The Borated Water Systems Stainless Steel Inspection is a one-time inspection activity for which there is no operating experience. However, an equivalent inspection was reviewed and deemed acceptable by the NRC Staff for Oconee, as stated in the conclusions below. Conclusion The Borated Water Systems Stainless Steel Inspection described above is equivalent to the Reactor Building Spray System Inspection described and evaluated in NUREG-1723, Section 3.5.3.2 [Reference B - 5]. Based on the above review, implementation of the Borated Water Systems Stainless Steel Inspection will adequately verify that no need exists to manage aging effects on the component or will otherwise take appropriate corrective actions so that the components will continue to perform their intended function(s) for the period of extended operation (i.e., 20-years from the end of the initial operating license). B.3.4-3

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B.3.5 BOTTOM-MOUNTED INSTRUMENTATION THIMBLE TUBE INSPECTION PROGRAM Note: The BOTTOM MOUNTED INSTRUMENTATION THIMBLE TUBE INSPECTION PROGRAM is generically applicable to both McGuire Nuclear Station and Catawba Nuclear Station, except as otherwise noted. The purpose of the Bottom Mounted Instrumentation Thimble Tube Inspection Program is to identify loss of material due to wear in the bottom mounted instrumentation (BMI) thimble tubes prior to leakage. The thimble tubes are part of the reactor coolant pressure boundary. The Bottom Mounted Instrumentation Thimble Tube Inspection Program is a condition monitoring program. Scope The scope of the Bottom Mounted Instrumentation Thimble Tube Inspection Program includes all thimble tubes installed in each reactor vessel. Preventive Actions No actions are taken as part of this program to prevent aging effects or mitigate aging degradation. Parameters Monitored or Inspected The Bottom Mounted Instrumentation Thimble Tube Inspection monitors tube wall degradation of the BMI thimble tubes. Failure of the thimble tubes would result in a breach of the reactor coolant pressure boundary. Detection of Aging Effects In accordance with information provided in Monitoring & Trending, the Bottom Mounted Instrumentation Thimble Tube Inspection Program will detect loss of material due to wear prior to component loss of intended function. Monitoring & Trending Inspection of the BMI thimble tubes is performed using eddy current testing. All of the thimble tubes are inspected. The frequency of examination is based on an analysis of the data obtained using wear rate relationships that are predicted based on Westinghouse research that is presented in WCAP-12866, Bottom Mounted Instrumentation Flux Thimble Wear [Reference B - 11]. These wear rates, as well as the results of the eddy current examinations are documented in site specific calculations. The eddy current results are trended and inspections are planned prior to the refueling outage in which thimble tube wear is predicted to exceeding the Acceptance Criteria, below. This ensures that the thimble tubes continue to perform their pressure boundary function. Acceptance Criteria The acceptance criteria for the BMI thimble tubes is 80% through wall (thimble tube wall thickness is not less than 20% of initial wall thickness). This acceptance criteria was developed by Westinghouse in WCAP 12866, Bottom Mounted Instrumentation Flux Thimble Wear, and reported to the NRC by Duke [Reference B - 10]. B.3.5-1

Corrective Action & Confirmation Process Thimble tubes that are predicted to exceed the acceptance criteria may be capped or repositioned. Specific corrective actions and confirmatory actions are implemented in accordance with the corrective action program. Administrative Controls Data are collected and evaluated using written procedures. The data are evaluated and the timing for the next inspection are determined using engineering calculations using methodology based on the information Westinghouse developed in WCAP-12866 [Reference B - 11]. Operating Experience Flux thimble wear was first identified as an issue when three flux thimble wore through over a three month period at the Salem plant in 1981. Since that time numerous plants both in the U. S. and abroad have detected thimble wear in varying degrees, ranging from small amounts to through wall. Westinghouse has determined the cause of this wear to be flow induced vibration of the flux thimble inside of the reactor vessel lower internals support column. Wear of the thimbles is a concern since the thimble serves as a portion of the reactor coolant system pressure boundary. On July 26, 1988, the NRC issued IE Bulletin 88-09: Thimble Tube Thinning in Westinghouse Reactors [Reference B - 12]. The NRC requested that inspection programs be implemented that included: The establishment, with technical justification, of an appropriate thimble tube wear acceptance criterion (for example, percent through wall loss). This acceptance criterion should include allowances for such items as inspection methodology and wear scar geometry uncertainties. The establishment, with technical justification, of an appropriate inspection frequency (for example, every refueling outage). The establishment of an inspection methodology that is capable of adequately detecting wear of the thimble tubes (such as eddy current testing). Duke has implemented a program at McGuire and Catawba Nuclear Stations that meets these criteria based on a proprietary study performed for the Westinghouse Owners Group [Reference B - 11]. Since the issuance of IE Bulletin 88-09 three inspections have been performed on Catawba Unit 1 and three on Catawba Unit 2 thimble tubes. The inspections on Unit 1 were performed during End Of Cycle (EOC) 1EOC-3 (1988), 1EOC-7 (1993) and 1EOC-11 (1999). The Inspections on Unit 2 were performed during 2EOC-2 (1989), 2EOC-3 (1990), and 2EOC-5 (1993). Inspections are not detecting significant changes in wear rates for either Unit. Currently no tubes are capped on Unit 1 and two tubes are capped on Unit 2 due to wear concerns. Wear projections performed in the referenced calculations have determined that B.3.5-2

further testing will not be required until 1EOC-7 (2008) and 2EOC-15 (2007), respectively for Units 1 and 2, barring significant changes in cycle length or reactor geometry. Similar inspections have been performed on McGuire Units 1 and 2. Unit 1 has been inspected twice, during 1EOC-5 (1988) and 1EOC-14 (2001) with 10 tubes showing detectable wall loss. Two additional tubes were capped due to other types of damage. Unit 2 was inspected during 2EOC-5 (1989) and 2EOC-8 (1993), with eight tubes showing wear. The future inspections are currently planned to occur at 1EOC-19 (2008) for Unit 1 and 2EOC-16 (2005) for Unit 2. Conclusion The Bottom Mounted Instrumentation Thimble Tube Inspection Program has been demonstrated to be capable of identifying loss of material due to wear in the thimble tubes prior to leakage. Based on the above review, the continued implementation of the Bottom Mounted Instrumentation Thimble Tube Inspection Program provides reasonable assurance that the aging effect will be managed and that the bottom mounted instrumentation will continue to perform its intended function for the period of extended operation (i.e., 20-years from the end of the initial operation license). B.3.5-3

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B.3.6 CHEMISTRY CONTROL PROGRAM Note: The Chemistry Control Program is generically applicable to both McGuire Nuclear Station and Catawba Nuclear Station, except as otherwise noted. The purpose of the Chemistry Control Program is to manage loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, and treated water environments. This program manages the relevant conditions that lead to the onset and propagation of loss of material and cracking which could lead to a loss of structure or component intended functions. Relevant conditions are specific parameters such as halogens, dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations that could lead to loss of material and/or cracking if not properly controlled. The Chemistry Control Program is a mitigation program. Scope The scope of the Chemistry Control Program is the mechanical components exposed to borated water, closed cooling water, fuel oil, and treated water environments in the following Catawba and McGuire systems: Auxiliary Feedwater System Auxiliary Steam System Auxiliary Ventilation System (CNS Only) Boron Recycle System Building Heating Water or Heating Water System Chemical and Volume Control System Component Cooling System Condensate System (CNS Only) Condensate Storage System (CNS Only) Containment Spray System Control Area Chilled Water System Control Area Ventilation or Control Room Area Ventilation Conventional Chemical Addition System (MNS Only) Diesel Generator Cooling Water System Diesel Generator Fuel Oil System Diesel Generator Lube Oil System Demineralized Water or Make-up Demineralized Water System Equipment Decontamination System (CNS Only) Feedwater System Feedwater Pump Turbine Exhaust System or Turbine Exhaust System Ice Condenser Refrigeration System Liquid Radwaste or Liquid Waste Recycle System Liquid Waste Monitor and Disposal System (MNS Only) Main Steam System Main Steam Supply to Auxiliary Equipment System Main Steam Vent to Atmosphere System Nuclear Sampling System Reactor Coolant System Recirculated Cooling Water System (CNS Only) Refueling Water System Residual Heat Removal System Safety Injection System Spent Fuel Cooling System Standby Shutdown Diesel System Steam Generator Blowdown or Steam Generator Blowdown Recycle System Steam Generator Wet Lay-up Recirculation System Waste Gas System The scope of the program also includes the spent fuel pool liner, structural stainless steel and plates, and racks located in the spent fuel pool. B.3.6-1

Preventive Actions The Chemistry Control Program monitors and controls the relevant conditions such as halogens, dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations to manage loss of material and cracking. These corrosive contaminants are either removed, their concentrations minimized, or treatments are added and/or maintained to negate their corrosive tendencies. Parameters Monitored or Inspected The Chemistry Control Program monitors specific parameters such as halogens, dissolved oxygen, conductivity, biological activity, and corrosion inhibitor concentrations. The specific parameters monitored vary depending on the system. Detection of Aging Effects No actions are taken as a part of this program to detect aging effects. Monitoring & Trending The Chemistry Control Program manages the following environments: (1) borated water, (2) closed cooling water, (3) fuel oil, and (4) treated water. For components exposed to borated water, the Chemistry Control Program draws and analyzes samples for contaminant concentrations to mitigate loss of material and/or cracking of components. Concentrations of contaminants such as dissolved oxygen, halogens, and sulfates are determined on a periodic basis. Monitoring and controlling the environment in the Reactor Coolant, Refueling Water, and Spent Fuel Cooling Systems will control the borated water environment and mitigate aging in the following license renewal systems. Boron Recycle Chemical and Volume Control Containment Spray Equipment Decontamination (CNS Only) Nuclear Sampling Residual Heat Removal Safety Injection Monitoring and trending in the Reactor Coolant, Refueling Water and Spent Fuel Cooling Systems ensures quick detection of unfavorable trends and prompt corrective actions to mitigate corrosion. For components exposed to closed cooling water, the Chemistry Control Program draws and analyzes samples for contaminant and treatment concentrations to mitigate loss of material and/or cracking of components. Concentrations of corrosion inhibitors are determined on a periodic basis to be within specific ranges. Monitoring and controlling the closed cooling water environment in the following systems will mitigate aging of components in these systems. B.3.6-2

Auxiliary Building Ventilation Building Heating or Heating Water Component Cooling Control Area Chilled Water Diesel Generator Cooling Water Ice Condenser Refrigeration Recirculated Cooling Water Standby Shutdown Diesel In addition, monitoring and controlling the closed cooling water environment in the above systems will mitigate aging of heat exchangers in the following systems exposed to the closed cooling water environment of the above systems. Chemical and Volume Control Control Area or Control Room Area Ventilation Diesel Generator Lube Oil Residual Heat Removal Waste Gas Monitoring and trending in the systems listed above ensures quick detection of unfavorable trends and prompt corrective actions to mitigate corrosion. For components exposed to fuel oil, the Chemistry Control Program draws and analyzes samples for contaminant concentrations to mitigate loss of material and/or cracking of components. Concentrations of contaminants such as water and biological activity are determined on a periodic basis. Monitoring and controlling the environment in the Diesel Generator Fuel Oil and Standby Shutdown Diesel Systems will control the fuel oil environment and mitigate aging. Monitoring and trending in these systems ensures quick detection of unfavorable trends and prompt corrective actions to mitigate corrosion. For components exposed to treated water, the Chemistry Control Program draws and analyzes samples for contaminant and treatment concentrations to mitigate loss of material and/or cracking of components. Concentrations of contaminants such as dissolved oxygen, halogens, and sulfates are determined on a periodic basis. Monitoring and controlling the treated water environment in the Demineralized Water, Feedwater, and Steam Generator Wet Lay-up Recirculation Systems will mitigate aging of components exposed to treated water in the following systems. B.3.6-3

Auxiliary Feedwater Auxiliary Steam Condensate (CNS Only) Condensate Storage (CNS Only) Conventional Chemical Addition (MNS Only) Equipment Decontamination (CNS Only) Feedwater Pump Turbine Exhaust or Turbine Exhaust Liquid Radwaste or Liquid Waste Recycle Liquid Waste Monitor and Disposal (MNS Only) Main Steam Main Steam Supply to Auxiliary Equipment Main Steam Vent to Atmosphere Nuclear Sampling Steam Generator Blowdown or Steam Generator Blowdown Recycle Steam Generator Wet Lay-up Recirculation Monitoring and trending in the Demineralized Water, Feedwater, and Steam Generator Wet Lay-up Recirculation Systems ensures quick detection of unfavorable trends and prompt corrective actions to mitigate corrosion. Acceptance Criteria The Chemistry Control Program contains system specific acceptance criteria that are based on the guidance provided in the EPRI chemistry guidelines [References B - 13, B - 14, and B - 15], Technical Specifications [References B - 16 and B - 17, Specification 3.8.3], UFSAR [References B - 18 and B - 19], and vendor recommendations for water and fuel oil quality. Corrective Action & Confirmation Process The Chemistry Control Program provides corrective actions when monitored parameters are trending unfavorably but have not violated an acceptance criteria. Additional sampling and analysis are performed after the corrective actions have been taken to confirm the effectiveness of the corrective actions in returning the parameters to acceptable levels. Parameters that have exceeded their acceptance criteria are entered into the corrective action program for a fuller investigation in addition to the corrective actions required by the Chemistry Control Program to return the parameter to acceptable levels. Specific corrective actions as a result of the fuller investigation are implemented in accordance with the corrective action program. Administrative Controls The Chemistry Control Program is controlled by the site program manuals and implemented by controlled plant procedures. The program manuals at each site provide guidance for maintaining a suitable system environment. These manuals are based on the guidance of the EPRI chemistry guidelines [References B - 13, B - 14, B - 15], Technical Specifications [References B - 16 and B - 17, Specification 3.8.3], the UFSAR [References B - 18 and B - 19] and vendor recommendations to manage loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, or treated water. Operating Experience A review of operating experience did not reveal a loss of the component intended function of components exposed to borated and treated water that could be attributed to the inadequacy of the Chemistry Control Program. This operating experience B.3.6-4

confirms the effectiveness of the Chemistry Control Program for borated and treated water to manage the aging effects when continued into the extended period of operation. Operating experience did reveal several instances of cracking at welds due to nitrate induced stress corrosion of carbon steel components in the Component Cooling Systems at Catawba and McGuire. A review of industry operating experience identified incidents of nitrate induced stress corrosion cracking of carbon steel components in comparable systems at several other utilities. The investigation determined that biological activity in protected areas were converting the nitrite corrosion inhibitors to nitrates, creating a highly corrosive localized environment. This occurred when nitrate concentrations were allowed to drift to higher than recommended limits. The Chemistry Control Program was modified to more rigorously control and lower system corrosion inhibitor concentrations along with the addition of biocides to control biological activity. The conductivity and conductivity to nitrite ratios are monitored as well as nitrate concentration. No other instances of nitrate induced stress corrosion cracking of carbon steel in the Component Cooling System have occurred since these changes were implemented. A review of operating experience did not reveal any instances of a loss of the component intended function of components exposed to fuel oil that could be attributed to the inadequacy of the Chemistry Control Program. Inspection of the Diesel Generator Fuel Oil System storage tanks revealed a light oxide layer and minor pits in the tank low point where water is likely to collect. The remaining internal surfaces of the tanks were free of aging degradation. These inspections and operating experience confirm the effectiveness of the Chemistry Control Program for fuel oil to manage the aging effects when continued into the extended period of operation. Conclusion The Chemistry Control Program has been demonstrated to be capable of managing loss of material and/or cracking of components exposed to borated water, closed cooling water, fuel oil, and treated water environments. The Chemistry Control Program described above is equivalent to the corresponding program described and evaluated in NUREG-1723, Section 3.2.2 [Reference B - 5]. Based on the above review, the continued implementation of the Chemistry Control Program provides reasonable assurance that the aging effects will be managed and that the components will continue to perform their intended function(s) for the extended period of operation. B.3.6-5